• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.581-592
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• 2008

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of $\gamma$-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.

• Journal of Advanced Engineering and Technology
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• v.4 no.4
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• pp.431-436
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• 2011

This study was carried out to reduce the problem during distillation process, which separate U, TRU (TRans Uranium) metal electro deposit, Cd and LiCl-KCl eutectic salt generating from LCC (Liquid Cadmium Cathode) electro winning process. The cadmium recovering apparatus was manufactured to separate for each metal using solid-liquid separation method. The apparatus consists of the first sieve for the separation of U and TRU metal electrodeposit, the second sieve for the separation of LiCl-KCl eutectic salt, cadmium collection basket, and a heating furnace. In addition, the size of each sieve is 2 mm to 3 mm. In this experiment, a metal wire was employed to replace TRU metal electrodeposit and U, which exist actually in a LCC crucible. In the solid state, The LiCl-KCl is separated at 340℃ at which the solid and the liquid of the remaining cadmium and LiCl-KCl eutectic salt coexists in each, after the metal wire separated at 500℃. As a result, it seems that it would be beneficial to set the processing condition in the distillation process with the additional treatment process of cadmium and LiCl-KCl eutectic salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.283-295
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• 2007

The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.0 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of ${\gamma}$-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.83-83
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• 2017

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. Distillation process was employed for the cathode processing. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. In this study, a mesh-covered crucible was investigated for the sat distillation of electrorefiner uranium deposits. A liquid salt separation step and a vacuum distillation step were combined for salt separation. The adhered salt in uranium deposits was efficiently removed in the mesh-covered crucible. The salt distiller was operated simply since repeated cooling - heating step was not necessary for the change of the crucible. The operation time could be reduced by the use of the mesh-covered crucible and the combined operation of the two steps. A method to preserve a vacuum level was proposed by double O-rings during the operation of the distiller with the mesh-covered crucible. After the salt distillation, the salt content was measured and was below 0.1wt% after the salt distillation. The residual salt after the salt distillation can be removed further during melting of uranium metal.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.74-74
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• 2017

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. Uranium deposit recovered from the solid cathode is a dendritic powder. It is necessary to separate the adhered salt from the deposits prior to the consolidation of uranium deposit. The adhered salt is composed of lithium, potassium, uranium, and rare earth chlorides. Distillation process was employed for the cathode processing. One of the operation methods is distillation of the salt at low temperature ($900^{\circ}C$), and then melting of the deposit at high temperature to avoid a backward reaction. For the development of the salt distiller, the distillation behavior of the low vapor pressure chlorides should be studied. Rare earth chlorides in the adhered salt of uranium deposits have relatively low vapor pressures compared to the process salt (LiCl-KCl). In this study, the evaporation rates of the lanthanum and neodymium chlorides were measured for the salt separation from electrorefiner uranium deposits in the temperature range of $825{\sim}910^{\circ}C$. The evaporation rate of both chlorides increased with an increasing templerature. The evaporation rate of lanthanum chloride varied from 0.12 to $1.68g/cm^2/h$. Neodymium chloride was more volatile than lanthanum chloride. The evaporation rate of neodymium chloride varied from 0.20 to $4.55g/cm^2/h$. The evaporation rate of both chlorides are more than $1g/cm^2/h$ at $900^{\circ}C$. Even though the evaporation rates of both chlorides were less than that of the process salt, the contents of the lanthanide chlorides were small in the adhered salt. Therefore it can be concluded that $900^{\circ}C$ is suitable for the operation temperature of the salt distiller.

• Analytical Science and Technology
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• v.17 no.4
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• pp.302-307
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• 2004

Determination of actinides in the environmental sample requires separation of each element. This procedure is tedious and time consuming. And also, the detection limits of some nuclides using alpha spectrometry are rather higher. To overcome the lower detection limit and complicated separation procedure, a simple analytical technique for the determination of actinide isotopes in the environmental samples was developed and applied to IAEA and NIST reference sediment samples. For the separation of actinides from matrix, anion exchange resin and TRU-spec extraction chromatography resin were used and chemical yields were obtained using natural uranium, thorium, $^{242}Pu$ and $^{243}Am$ tracers. For overcoming the higher detection limits of U and Th in alpha spectrometry, neutron activation analysis was applied. Using combined method, the detection limit was increased about 10 times. The activity values of each isotope were consistent with the reference values reported by IAEA and NIST.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.351-357
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• 2003

Transuranic elements such as Pu, Am and Cm in synthetic solution of spent nuclear fuel samples were determined by electrodeposition followed by alpha-spectrometry after separation using anion exchange and extraction chromatography in order to determine the transuranic elements in radwaste samples from nuclear power plants. Plutonium was separated by 12M HC1-0.1M HI as an eluent on anion exchange column. As a second step Am and Cm were separated in a group by DTPA-Lactic acid as the eluent on HDEHP coated column. The nuclides of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ separated were determined by alpha-spectrometry after electrodeposition in 0.1M $NaHSo_4$-0.53M $Na_2SO_4$buffer solution as an electrolyte. The recovery yields of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ were 83.8%, 85.2% and 86.3%, respectively, from the synthetic solution containing uranium and non-radioactive metal elements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.59-64
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• 2018

In case of high contents of rare earths in the LiCl-KCl salt, it is not easy to recover U and TRU metals as a usable resource form from LiCl-KCl eutectic salts generated from the pyroprocessing of spent nuclear fuel. In this study, a conversion of $UCl_3$ into an oxide form using $K_2CO_3$ and an electrodeposition of $NdCl_3$ into a metal form in $LiCl-KCl-UCl_3-NdCl_3$ system were conducted to resolve the problem. Before conducting the conversion, experimental conditions for the conversion were determined by performing a thermodynamic equilibrium calculation. In this study, almost all of $UCl_3$ disappeared in the LiCl-KCl salt when the injection of $K_2CO_3$ reached theoretical equivalent for the conversion, and then $NdCl_3$ was effectively electrodeposited as a metal form using liquid zinc cathode. After that, the LiCl-KCl salt became transparent, and uranium oxides were precipitated to the bottom of the LiCl-KCl salt. These results will be utilized in designing a process to separate U and rare earths in LiCl-KCl salt.